Thermohydraulic Code Research Articles

SIMULATION OF TRANSIENT PROCESS ON WWER-1000 USING COUPLED APPROACH

Research and modeling of thermohydraulic processes in NPP equipment is necessary for a deep and improved justification of reliable and safe operation of nuclear installations. Taking into account the accumulated experience and the latest technologies, the study of thermohydraulic processes is developing in the direction of applying innovative approaches to conducting both experimental studies and expanding the calculation capabilities of thermohydraulic codes. The article presents an approach to the decomposition of the thermohydraulic model of the WWER-1000 reactor plant on submodels, which are calculated in coupling with separate instances of the system thermohydraulic code RELAP5/Mod 3.2. Coupling is implemented using a specially developed module. This approach allows you to "bypass" the internal limitations of RELAP5 in the number of components (hydrodynamic volumes, thermal structures, "trips" and control variables) that can be used for computational analysis. As a result, it becomes possible to achieve a higher degree of detail for the entire reactor installation. In addition, the article confirms the possibility of applying the previously developed coupling module to such a non-equilibrium model as a nuclear reactor model. At the same time, the use of the RELAP5 model as a reactor model makes it possible to avoid inaccuracies that arise when applying assumptions about the transverse profiles of the input variables of the CFX model, as well as to perform the validation of the RELAP5 model of the RU loops. Coupled calculation of stationary and transient mode of reactor installation WWER-1000 is performed. The reliability of calculation results obtained in conjugation is evaluated by comparing them with the results of autonomous calculations in RELAP5. The analysis of the results of computational modeling confirms the possibility of applying the developed coupling module to such a non-equilibrium model as a model of a nuclear reactor, and also confirms the correctness of the RELAP5 model of the loops of the reactor plant obtained on the basis of the modification of the full-fledged RELAP5 model.

Subdiffusive processes in BWRs

The objective of this work is the analysis of subdiffusive processes in BWR reactors. Nuclear reactors are highly heterogeneous systems where the phenomena at the reactor scale are not possible to describe with constitutive standard laws (normal diffusion) of energy, momentum, and mass. However, there are methods to achieve it such as the volumetric average to upscale the reactor and incorporate up-scaled constitutive laws. However, it can also be proposed constitutive laws of fractional order to consider anomalous diffusion. For the neutronic processes, the neutron diffusion theory is non-appropriated to the reactor scale, because this theory is applied to the homogeneous reactor. Given the heterogeneity of these systems in describing the dynamic behaviour of BWRs with point models, it represents a challenge to the incorporation of the heterogeneous effects whose main characteristic is that the transport phenomena are subdiffusive with relaxation times (i.e., non-instantaneous) by nature. The point models called reducer-order models are important for stability analysis and to develop control strategies. In linear or nonlinear stability analysis and system control, they are carried out with reduced order models that are zero dimension, i.e., point models that depend exclusively on time. This is because existing methodologies propose mathematical expressions in the time or frequency domain for stability analysis or to establish control strategies. The challenge is for ROM models to be able to consider subdiffusive effects on neutron motion. This is achieved by various routes from the P1 theory that includes the temporal term with relaxation effects (relaxation time) or by proposing a more general constitutive equation of fractional order for the neutron current vector, as is presented in this work. Moreover, the Fractional Order Models behavior must be as a transport theory based model, as well as in those nuclear and thermohydraulic codes which are very useful but complex, for which is sometimes it is impossible to obtain analytical expressions. To incorporate these effects in this work we consider point models of fractional order, which represent the subdiffusive phenomenon when the fractional order (known as anomalous diffusion coefficient) is greater than 0 and less than 1.

Extension of the X2 VVER-1000 benchmark by a control rod cluster ejection exercise

This paper proposes a new benchmark for VVER-1000 control rod ejection transient. The benchmark is designed for code-to-code comparison and its purpose is testing and verifications of time-dependent solvers, core models and cross section generation methodology. Four solutions obtained using different combinations of three lattice transport, two nodal neutronics, and two thermo-hydraulic codes are analyzed, demonstrating good agreement in predicting transient reactivity and power peaks and highlighting the impact of gas gap modeling and DNBR correlation on simulation results.

Modeling hot channel two-phase flow of PWR NPPs during abnormal condition, using Python

The boiling and equilibrium characteristics of subcooled region would be considered as an important advantage in PWRs from the Thermal-Hydraulic point of view. However, the temperature gradient at some sections of reactor core and high temperature of clad surface lead to two–phase condition especially in transient conditions. The high temperature cause the heat transfer regime to change while observing the micro-vapor bubbles in some parts of the fuel channels, and a part of the heat flux changes the liquid phase to steam. This change requires the use of new models to predict fluid/wall temperature and non-equilibrium thermodynamic characteristics. In this research, the simulating of fuel channels during normal and abnormal conditions (PLOFA) was performed by employment of heat transfer models predicting the equilibrium and non-equilibrium characteristics based on the thermohydraulic code developed in Python. The comparison of the results (e.g. flow pattern and temperature profile) and the experimental data shows a good agreement for two-phase flow.

RELAP5-CFX COUPLING MODULE: TESTING ON SIMPLE THERMOHYDRAULIC MODELS

The modern direction in the study of the safety of nuclear power plants is to ensure the maximum level of detail in process modeling with a satisfactory level of computational resources. One of the approaches of such task solving is the coupled use of special software with the required levels of detail, for example, systemic thermohydraulic codes with computational hydrodynamics codes.
 This article describes the developed coupling module between the system thermohydraulic code RELAP5/Mod. 3.2 and the CFX computational hydrodynamics code, using an external explicit interfacing scheme without overlap, in which CFX acts as the main code and RELAP5 as the secondary one. The development of this module is part of the work on integrated coupled modeling of severe accidents and makes it possible to model the defining processes and phenomena that occur during a severe accident within the reactor vessel using the CFD model, and the elements of the reactor unit using the RELAP5 model with an acceptable cost of resources and with a high level of detail. To check the coupling technology, the module was tested on simplified models containing serially connected RELAP5- and CFD-models of horizontal pipes, as well as on a closed loop. In view of the further use of the interface module on a full-fledged model of the VVER-1000, testing is performed for three main interface types: "input flow", "output flow" and their combination "input/output flow" on a closed loop. After completion of the testing, the accuracy and credibility of the results of the coupled calculations was evaluated by comparing them with the results of autonomous RELAP5 calculations. Results of the tests calculations support the possibility of using the RELAP5-CFX coupling module on more complex and large-scale RELAP5 and CFD models.

Proposal of point-wise power reconstruction for IRT-4M fuel assemblies in coupled Monte-Carlo calculations

Current efforts to develop high-fidelity multi-physics tools for research reactors with IRT-4M fuel leads to the involvement of Monte-Carlo transport and subchannel thermo-hydraulic codes in the process of safety assessment. Monte-Carlo based neutronic codes are often used as a reference tool and are suitable for a prediction of a detailed power distribution. This paper proposes a method of the point-wise power reconstruction based on a Radial Basis Function (RBF) interpolation for complicated irregular geometries and compares different detector functionalities of the Serpent2 code from the time efficiency point of view. Furthermore, a preliminary validation of the applied method for IRT-4M in a subchannel code SUBSALS mesh is presented. This code is currently under the evelopment at the Research Centre Rež and the Czech Technical University in Prague and focuses on the thermohydraulic analysis of IRT type fuel assemblies.

Three-Liquid, Two-Phase, Per-Channel Thermohydraulics Code Subchannel-Na/V1.0 for Single-Phase Sodium-Coolant Flows and Heat-Transfer: Key Models

Three-Liquid, Two-Phase, Per-Channel Thermohydraulics Code Subchannel-Na/V1.0 for Single-Phase Sodium-Coolant Flows and Heat-Transfer: Key Models

PROBLEMS OF MODELING IN-VESSEL SEVERE ACCIDENT PHASE

The process of a severe accident at a nuclear power plant is characterized by its complexity, versatility and wide relationships. When modeling the in-vessel phase of a severe accident, there are significant uncertainties, primarily related to the movement of materials of the destroyed core to the lower part of the reactor vessel, as well as its subsequent heating and accompanying high-temperature physical processes. A detailed comprehensive study of these processes will allow us to reliably predict the dynamics of a severe accident and correctly determine the time of failure of the reactor vessel and justify more effective actions of personnel in managing severe accidents and emergency response.
 The article presents an analysis of the course of a severe accident, identifies and analyzes the main phases of the in-vessel phase of a severe accident. The analysis of processes and phenomena characteristic of these phases is carried out. The main uncertainties that arise when modeling the processes of a severe accident within the reactor vessel are considered. To eliminate/reduce uncertainty, an approach to conjugate severe accident modeling using the ANSYS CFX CFD code and the RELAP5/Mod 3.2 system thermohydraulic code is proposed.

Dynamic simulation on flow characteristics of KSTAR PF magnet cryogenic network

During its nominal operation, strong current variation, more than 10 kA per second, can be imposed on the KSTAR PF (Poloidal Field) magnet system. Due to AC loss, transient massive backward flow can occur, especially at the cryogen inlets of each magnet. To protect cryogen circulator, several valves are activated to adjust or lower pressure below the circulator's operation limit. In this work, we discuss whether currently available thermo-hydraulic code, such as SUPERMAGNET, can describe this overall dynamic flow characteristics of KSTAR PF magnet cryogenic network. First, three major functions of the KSTAR PF magnet cryogenic system during plasma operation are classified and discussed. Cryogenic components mostly related with other functions, such as cool-down process, are removed from the simulation circuit to minimize computation time. SUPERMAGNET code was used with a slight modification in its cryogenic network simulation module, FLOWER. The codes for cryogenic valves are modified so that not only steady state but also general compressible flows can be described. A case study for an actual KSTAR plasma operation has been intensively carried out. It was shown that most relevant data, pressure, temperature and mass-flow, at the inlets and outlets of the magnet, are in agreement with simulation results.

Development and quality assurance of computer programs for the review of long-term safety analysis

Abstract. During the site selection process, regulated by the Site Selection Act (Standortauswahlgesetz – StandAG), the implementer has to identify adequate siting regions and has to perform long-term safety analysis for these regions. The Federal Office for the Safety of Nuclear Waste Management (Bundesamt für die Sicherheit der nuklearen Entsorgung – BASE) as the responsible federal authority has to review the implementer's long-term safety analysis. To perform this duty in the required depth, it will be necessary to recalculate important aspects of the analysis by means of numerical computer programs. In addition, this will allow assessment of the underlying uncertainties of the implementer's long-term safety analysis from a regulatory point of view. Numerical modelling requires a high degree of quality assurance. Therefore, it is of vital importance that the same problem is modelled with different computer programs and – if possible – by different teams of modelers. This strategy is known as the diverse modelling approach. The goal of this approach is to perform a cross-check of the computer programs in use as well as of the correctness of data handling, data interpretation, and model implementation. Comparing outcomes of different modelling teams also integrates complementary views and approaches, which can be beneficial when dealing with complicated problems. The diverse modelling approach forms the basis of regulatory modelling. It can also be implemented within organizations if more than one code is used to tackle the same task. The diverse modelling approach has been carried out at GRS gGmbH over the past two decades by means of the computer programs, TOUGH2-GRS and MARNIE(2). Both (co-)developed programs are thermohydraulic codes that compute transport phenomena in porous media and can be coupled to geochemical codes, see Navarro (2018) and Navarro et al. (2021) for further details. TOUGH2-GRS and MARNIE(2) were applied in various projects, such as preliminary safety analysis Gorleben (VSG, Vorläufige Sicherheitsanalyse Gorleben), ZIESEL and EVEREST. A recent example for the diverse modelling approach was the use of TOUGH2-GRS and MARNIE for the development of indicators for the safe confinement of radionuclides in a deep geological repository (Navarro et al., 2019). A high degree of quality assurance is also achieved by involving computer programs in benchmarks to compare results among different codes for well-defined problems. For the calculation of THM (thermo-hydraulical-mechanical) processes TOUGH2-GRS has been coupled with the geomechanical computer program FLAC3D and this approach has been used in the benchmark BenVaSim (Seher et al., 2019). In this contribution selected works related to TOUGH2-GRS and MARNIE(2) are presented. From this starting-point, ideas and concepts for the continuation of development and quality assurance of the two computer programs at BASE in the near future are outlined. These works will contribute to strengthen the capabilities of BASE in the independent review process of implementer's long-term safety analysis within the site selection process.

Investigation of UO2 fuel efficiency for ABV small modular reactor

ABSTRACT In recent years, small modular reactors have received considerable attention in various countries due to having some capabilities such as small dimensions. The ABV reactor is a small modular reactor with a power of 45 MWth. For this reactor, various applications such as use as a desalination plant, source of electrical and thermal energy, etc. have been considered. In this reactor, an unconventional metal fuel called silumin with 16.5% enrichment and a cycle length of 3000 days has been used. The main purpose of this research is to investigate the performance of conventional UO2 fuel while maintaining the initial cycle length as the main fuel of ABV reactor. To achieve this purpose, the performance of UO2 fuel with enriches of 8%, 10%, 12%, 14%, and 16% with different percentages of Gd2O3 has been investigated. To perform the calculations, the RELAP5 thermohydraulic code and the PARCS neutronic code have been coupled by using a Parallel Virtual Machine, and the reactor core quantities included: effective multiplication factor, relative power distribution, feedback coefficients, plutonium and uranium isotope concentrations, axial and radial temperature distribution of fuel rod, heat flux and DNBR have been calculated and is within an acceptable range for this design.

Heat-Exchange Regularities and Thermohydraulic Code Verification for Supercritical Pressure Water Cooled Reactors

Heat-Exchange Regularities and Thermohydraulic Code Verification for Supercritical Pressure Water Cooled Reactors

ГИДРОДИНАМИКА ТУРБУЛЕНТНЫХ ПОТОКОВ В ТВС БЫСТРЫХ РЕАКТОРОВ (ПОЛЕ СКОРОСТИ И МИКРОСТРУКТУРА ТУРБУЛЕНТНОСТИ)

The article describes the factors under the influence of which the formation of thermohydraulic characteristics occurs in the fuel assemblies of the core of fast reactors with liquid metal cooling. It is shown that one of the most important factors is a complex multiply connected geometry of a stochastic nature, subject to deformation during the campaign under the influence of temperature irregularities and radiation effects. The paper presents and analyzes the results of experimental and computational studies of the velocity field and shear stress, the microstructure of turbulence, momentum transfer in the central and peripheral regions of fuel assemblies without and with displacers, as well as in the case of deformation of the lattice of rods. The intensification of turbulent momentum transfer in the channels in the azimuthal direction in the area of the gaps between the rods is demonstrated. The anisotropy coefficient of turbulent momentum transfer reaches 30-40 units. The performed analysis indicated a significant difference in the calculated in the framework of semi-empirical models of turbulent transfer and experimental dependences of the coefficients of turbulent transfer of momentum in the radial and azimuthal directions and the coefficients of anisotropy of turbulent transfer of momentum in rod bundles. The results of an open benchmark on the thermohydraulics of fuel assemblies showed that common commercial computational thermohydraulic codes only approximately describe the experimental data. It is shown that the intensification of turbulent momentum transfer in the channels of rod assemblies is due to the appearance of large-scale turbulent momentum transfer (secondary flows). The contribution of large-scale turbulent momentum transfer to the kinetic energy of turbulent pulsations, azimuthal turbulent shear stresses, and turbulent momentum transfer coefficients in rod assemblies is calculated. An empirical dependence of the coefficient of interchannel turbulent impulse exchange in bundles of smooth rods is obtained, on the basis of a semi-empirical model, data on interchannel turbulent impulse exchange in assemblies of smooth rods are generalized, and the intensification of interchannel turbulent exchange in close lattices of rods is explained. Data on hydraulic resistance in bundles of smooth rods are analyzed. The tasks of further research are discussed.

РАСЧЕТ МОЩНОСТИ ПРЯМОТРУБНОГО ПАРОГЕНЕРАТОРА С ЖИДКОМЕТАЛЛИЧЕСКИМ ТЕПЛОНОСИТЕЛЕМ

The article investigates the influence of coolant flow profile nonlinearity through straight-pipe steam generators tube assembly for fast neutron reactor with sodium coolant on heat exchanger power. When designing steam generators, a very important task is to correctly calculate the output parameters of the steam generator, especially the power. For the plants without reference solutions it is necessary to perform a deep analysis of the factors affecting on calculated parameters and to incorporate these parameters in the codes. An example of a new plant that do not have analogues is a shell-type steam generator for perspective fast neutron plants with liquid metal sodium coolant. The application of new solutions in steam generators design requires experimental and calculational justification of thermal hydraulic with the use of modern calculation codes. Power calculation of steam generator is carried out by thermohydraulic code “KORSAR/GP”, “PGN-2K”. One of the assumptions in coolant path parameters calculation model (tube space) is a uniform velocity profile by cross-section of tube assembly. It’s also accepted, that each heat exchange tube has the same expense of feed water. On the other hand calculational CFD and experimental studies at aerodynamic model of steam generator showed the presence of significant unevenness of coolant expense by tube space cross-section which is not taken into account in thermohydraulic calculations. The article contains the methodic of accounting for the known uneven coolant flow profile by tube space cross-section in liquid metal steam generator calculation. Based on the results obtained, measures to improve power output and reliability are proposed.

ОСОБЕННОСТИ РАСЧЕТНОГО МОДЕЛИРОВАНИЯ ТЕПЛОВЫХ РЕЖИМОВ БАССЕЙНОВ ВЫДЕРЖКИ БиАЭС

The paper considers methodological aspects during the development of thermo-hydro-dynamic of numerical calculation models for spent nuclear fuel pool (SNFP) on the example of Bilibino NPP by using international industry codes. The purpose of these models development is the fast simulation of thermal and humidity operational modes of spent fuel pool during periods of filling, "wet" and dry storage with an interval of up to 50 years, taking into account passive and forced convection heat removal systems. The following methodological aspects are considered in detail: 1. Use of isotope kinetics codes for calculating of the heat power dynamics for separate spent fuel assembly. 2. Calculation method of the heat removal power from the evaporation mirror of SNFP during “wet” storage, including evaporation power calculation and exhaust ventilation operation. Using the law of similarity of heat transfer and mass transfer processes (Lewis' law) for evaporation calculations. 3. Methods of accelerated computational forecasting of the dynamics of the thermal regimes of the SNFP during “wet” storage. 4. Condensation of atmospheric moisture at the bottom of the SNFP after “dry” storage and methods for its removal. It is shown that TRAC (TRACE) code with a 3D porous body model and complete evaporation-condensation models is the most suitable for solving the problems under consideration among the system thermo-hydraulic two-phase codes for nuclear energy.

Prediction of PbLi fluid flow and temperature field in a thermal convection loop for qualification of fusion materials

Two modeling tools, a thermohydraulics code and a computational model in COMSOL Multiphysics, have been developed, tested and then applied to the analysis of fluid flow and heat transfer in a thermal convection loop (TCL). Such a device has recently been used to experimentally evaluate corrosion compatibility of APMT (Fe–21Cr–5Al–3Mo) steel with high-temperature molten eutectic lead-lithium (PbLi) alloy at Oak Ridge National Lab, TN, USA. The 1-D thermohydraulics code allows for rapid calculations of the loop parameters as a function of the applied heat flux. The supplemental COMSOL finite-element computations provide detailed 3-D velocity and temperature field data but are much more time consuming. Both modelling tools demonstrate an excellent agreement in the computed circulation velocity as well as maximum and minimum temperatures. The performed TCL analysis focuses on flow development in the “hot” and “cold” legs, formation of Dean vortices in corner regions, blocking effect of the immersed samples, and radiative and convective cooling effects under the experiment-relevant conditions for Prandtl number Pr=0.015, Grashof numbers Gr∼107, and Reynolds numbersRe∼104.

Simulation of the Super Critical Water Loop Using ATHLET Code During an Abnormal Scenario

Abstract Supercritical water (SCW) has advantages like high thermal efficiency and can operate at high temperature and pressure. At the same time, however, these properties bring up related issues, such as material compatibility and corrosion resistance. In an effort to fully investigate the operating conditions, and solutions to these issues, test facilities are being built by many research organizations. One such organization, the Research Center Řež (CVR) located in the Czech Republic, has developed an experimental supercritical water loop (SCWL). The purpose of this loop is to provide experimental data from material testing in various conditions, including operating under the neutron field. This will be achieved by inserting a test channel into the existing experimental reactor light water reactor 15 (LVR-15), which will require a license from the state nuclear regulator (State Office for Nuclear Safety (SUJB)). Part of the licensing documentation is the safety analysis, which combines results from developed models using the thermohydraulic code ATHLET 3.1 A patch 1, as well as the experimental out of pile data. Among the postulated scenarios, an abnormal sequence (labeled A2—Loss of power in the loop) was analyzed in order to provide a preliminary benchmark. This scenario is similar to the postulated in-pile A2 and it was used for the benchmark activity. The aim of this paper is to present this activity including the adopted assumptions in the model. In particular, the paper presents, how these assumptions influenced the results indicating the discrepancies obtained in the first part of the transient.

РАСЧЁТНЫЙ АНАЛИЗ ТЕПЛОГИДРАВЛИЧЕСКОЙ УСТОЙЧИВОСТИ ОБРАТНОГО ПАРОГЕНЕРАТОРА ИССЛЕДОВАТЕЛЬСКОЙ ЯДЕРНОЙ УСТАНОВКИ МБИР

The stability of the circulation in the steam-water circuit, that is, the preservation of the thermohydraulic parameters within acceptable limits for perturbations characteristic of the operation, largely determines the reliability of the steam generator itself. Under certain conditions, pulsating pressure, flow and temperature pulsations may appear in boiling apparatus. For the first time, ripples were encountered during the creation of once-through boilers, and the harmful effect of flow instability on the apparatus design led to the need for a detailed study of the problem. In this work, a computational analysis of the thermal-hydraulic stability of the reverse steam generator (OPG) of the MBIR nuclear research facility (NRF) is carried out. The need for such a study is due to the fact that the OPG consists of three parallel connected steam generating modules. With such a design, water flow rate pulsations may appear, including with overturning flow rates in separate steam-generating channels. To avoid this, it is necessary to conduct throttling at the inlet to each OPG module, therefore, in this study, the hydraulic resistance of the chokes is determined. The research methodology is based on the use of a numerical thermohydraulic model of a reverse steam generator, made by means of a certified thermohydraulic design code HYDRA-IBRAE/LM/V1. The OPG RNU MBIR at the economizer and superheating sections has twisted heat exchange intensifiers that direct water and steam both in the longitudinal and transverse directions relative to the bundle of heat exchange tubes. This circumstance required a special approach in calculating the hydraulic resistance and heat transfer coefficients along the path of the OPG modules. After the creation of the computational model, it had to be verified, and since the design of the OPG of the NRF MBIR is similar to the design of the OPG of the BOR-60 reactor, the experimental data obtained on the OPG BOR-60 were used for verification. The calculations of the normal operation modes of the OPG NRF MBIR after the throttling of the modules confirmed the absence of fluctuations in the flow rate in these modes.

SKETCH-N/ATHLET coupled calculations using the boundary conditions plugin

The paper describes recent developments in the multi-physics coupling scheme between the SKETCH-N nodal neutronics code and the best-estimate thermohydraulic code ATHLET v3.2. The boundary conditions plugin was implemented. The verification and validation were performed using the transient of the Kalinin-3 international Benchmark. The simulation results using the boundary conditions plugin show good agreement with experimental data and calculations performed by using SKETCH-N/ATHLET direct calculations. The calculations of the transient “Switch off of one MCP (Main Coolant Pump) at nominal power” is performed applying a simple core thermohydraulic model without taking into account inter-channel mass transfer.

Influence of cross-section uncertainties on the calculated results from a thermo-hydraulics code

This study deals with the influence of uncertainties in selected cross-section data in a neutronics code while their impact on the results from thermo-hydraulics code is investigated. Particular cross-section data were defined according to the Phenomenon Identification and Ranking Tables (PIRT) for a pressurized water reactor. Uncertainties were identified for the following data: β – delayed neutron fraction precursors and transport, absorption, ν-fission, κ-fission, down scattering for moderator temperature, moderator density and fuel temperature. As a neutronics code, PARCS in conjunction with thermo-hydraulics code TRACE was applied. Uncertainty analyses were carried out by the code DAKOTA where coupled PARCS/TRACE calculations were run 146 times in order to meet a two-sided 95/95 confidence interval. In the thermo-hydraulic code thermal power of the reactor core and inner temperature of the fuel rod were studied. This research shows that the most influential parameters associated with cross-section parametrization are moderator temperature and fuel temperature.